This invention relates to annealing members formed from Zircaloy 2 or Zircaloy 4 alloys to reduce the susceptibility of the member to nodular corrosion.
Nuclear fuel element cladding serves several purposes and two primary purposes are: first, to prevent contact and chemical reactions between the nuclear fuel and the coolant or the moderator if a moderator is present; and second, to prevent the radioactive fission products, some of which are gases, from being released from the fuel into the coolant or the moderator. The failure of the cladding, i.e., a loss of the leak-proof seal, can contaminate the coolant or moderator and the associated systems with radioactive long-lived products to a degree which interferes with plant operation.
Zirconium-based alloys have long been used in the cladding of fuel elements in nuclear reactors. A desirable combination is found in zirconium by virtue of its low thermal neutron cross-section and its generally acceptable level of resistance to corrosion in a boiling water reactor environment. Zircaloy 2, a zirconium alloy consisting of about 1.2 to 1.7 percent tin, 0.07 to 0.2 percent iron, 0.05 to 0.15 percent chromium, 0.03 to 0.08 percent nickel, up to 0.15 percent oxygen, and the balance zirconium, has been used in reactor service, but possesses some deficiencies that have prompted further research to improve performance. Zircaloy 4 was one alloy developed as a result of further research to improve Zircaloy 2. Zircaloy 4 is similar to Zircaloy 2 but contains less nickel (0.007% max. wt. percent) and slightly more iron. Zircaloy 4 was developed as an improvement over Zircaloy 2 to reduce absorption of hydrogen in Zircaloy 2. Zircaloy 2 and Zircaloy 4 are herein referred to as the Zircaloy alloys or Zircaloy. The Zircaloy 2 and Zircaloy 4 alloys are disclosed in U.S. Pat. Nos. 2,772,964 and 3,148,055, both incorporated herein by reference.
The Zircaloy alloys are among the best corrosion resistant materials when tested in water at reactor operating temperatures, typically about 290.degree. C., but in the absence of radiation from the nuclear fission reaction. The corrosion rate in water at 290.degree. C. is very low and the corrosion product is a uniform, tightly adherent, black ZrO.sub.2 film. In actual service, however, the Zircaloy is irradiated and is also exposed to radiolysis products present in reactor water. The corrosion resistance properties of Zircaloy deteriorate under these conditions and the corrosion rate thereof is accelerated.
The deterioration under actual reactor conditions of the corrosion resistance properties of Zircaloy is not manifested in merely an increased uniform rate of corrosion. Rather, in addition to the black ZrO.sub.2 layer formed, a localized, or nodular corrosion phenomenon has been observed in some instances on Zircaloy tubing in boiling water reactors. In addition to producing an accelerated rate of corrosion, the corrosion product of the nodular corrosion reaction is a highly undesirable white ZrO.sub.2 bloom which is less adherent and lower in density than the black ZrO.sub.2 layer.
The increased rate of corrosion caused by the nodular corrosion reaction will be likely to shorten the service life of the tube cladding, and also this nodular corrosion will have a detrimental effect on the efficient operation of the reactor. The white ZrO.sub.2, being less adherent, may be prone to spalling or flaking away from the tube into the reactor water. On the other hand, if the nodular corrosion product does not spall away, a decrease in heat transfer efficiency through the tube into the water is created when the nodular corrosion proliferates and the less dense white ZrO.sub.2 covers all or a large portion of a tube.
Actual reactor conditions cannot be readily duplicated for normal laboratory research due to the impracticality of employing a radiation source to simulate the irradiation experienced in a reactor. Additionally, gaining data from actual use in reactor service is an extremely time consuming process. For this reason, there is no conclusory evidence in the prior art which explains the exact corrosion mechanism which produces the nodular corrosion. This limits, to some degree, the capability to ascertain whether new thermal or mechanical treatments of members formed from Zircaloy will be susceptible to nodular corrosion before actually placing the members into reactors.
Laboratory tests conducted under the conditions normally experienced in a reactor at approximately 300.degree. C. and 1000 psig in water, but absent radiation, will not produce a nodular corrosion product on Zircaloy alloys like that found in some instances on Zircaloy alloys which have been used in reactor service. However, if steam is used with the temperature increased to over 500.degree. C. and the pressure raised to 1500 psig, a nodular corrosion product can be produced on Zircaloy alloy samples in laboratory tests. Such testing in steam at 500.degree. C. and 1500 psig to as the high-pressure steam test.
Research efforts directed at improving the corrosion properties of Zircaloy have yielded some advances. Corrosion resistance has been enhanced in some instances through carefully controlled heat treatments of the alloys either prior to or subsequent to material fabrication. For example, it was found that a high cooling rate from the beta or alpha-plus-beta range provides what is known as a beta-quenched crystal structure having good nodular corrosion resistance in the high-pressure steam test. Subsequent hot working or alpha annealing, such as recovery, partial recrystallization, or full recrystallization annealing after cold working decrease the nodular corrosion resistance of the beta-quenched structure.
It is known that improved nodular corrosion resistance is obtained when Zircaloy has been cold worked or quenched from the beta or alpha-plus-beta range, but the cold worked or beta-quenched structures are detrimental to other properties such as ductility, creep resistance, and toughness. A compromise to obtain mechanical properties and corrosion resistance is provided with the beta-quench prior to the final cold rolling and anneal. U.S. Pat. Nos. 4,450,016 and 4,450,020 disclose Zircaloy fuel cladding tubes formed by beta-quenching prior to a cold rolling, after which an anneal is performed at a temperature of 500.degree. to 610.degree. C. in vacuum. The cumulative time and temperature of each successive anneal after the beta-quench improves the creep and the uniform corrosion resistance, but unfortunately decreases the nodular corrosion resistance in the high-pressure steam test, see "Influence of Variations in Early Fabrication Steps on Corrosion, Mechanical Properties, and Structure of Zircaloy-4 Products," D, Charquet, E. Steinberg, Y. Miller, Zirconium in the Nuclear Industry: Seventh International Symposium, ASTM STP 939, American Society for Testing and Materials, 1987, pp 431-447.
For example, Charquet et al. disclose a cumulative annealing parameter that is a function of annealing time, temperature, and an emperically determined activation energy. FIG. 1, reproduced from the Charquet et al. disclosure, shows that as the annealing parameter increases for fully recrystallized material, the resistance to nodular corrosion substantially decreases. Zircaloy in the cold worked or as pilgered condition maintains a high resistance to nodular corrosion; however, the mechanical properties are not suitable for use as cladding for nuclear reactor fuel. The cold worked Zircaloy must be annealed to recover, partially recrystallize, or fully recrystallize the material to achieve the desired mechanical properties.
It is an object of this invention to provide a method for mitigating the reduction in nodular corrosion resistance of Zircaloy alloy members that are annealed.